The containment vessel failure mode, “molten fuel-coolant interaction outside the reactor pressure vessel” (“ex-vessel FCI”) is a phenomenon of rapid increase in the pressure inside the containment vessel or steam explosion caused by the contact between the molten reactor core and cooling water outside the reactor pressure vessel after the reactor core is damaged.

To evaluate the viability of keeping confinement function of the containment vessels of Units 6 and 7 (ABWRs) of Kashiwazaki-Kariwa Nuclear Power Station against “ex-vessel FCI,” we conducted a code-based event progression analysis. For evaluation of the rapid increase, we employed the severe accident analysis code, MAAP, after organizing the critical phenomena of the event. In addition, assuming a case of steam explosion occurrence, we conducted an analysis with employing the steam explosion analysis code, JASMINE, and the structural response analysis code, AUTODYN-2D.

As a result of the evaluation, the maximum pressure and temperature inside the containment vessels were lower than their limits. Moreover, the maximum stress applied to the lower part of the containment vessels was lower than the yield stress on support structure of the lower part of the containment vessels. Therefore, we could confirm that the containment vessels can keep their integrity.

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